1. Field of the Invention
The present invention relates to safety systems for nuclear reactors. More specifically, this invention is directed to the prediction of internal reactor conditions commensurate with maintaining the integrity of the fuel element cladding. Accordingly, the general objects of the present invention are to provide novel and improved apparatus and methods of such character.
2. Description of the Prior Art
The performance of a nuclear reactor, like that of many other energy conversion devices, is limited by the temperature which component materials will tolerate without failure. In the case of a reactor with a core comprising an assemblage of fuel assemblies which in turn consist of an array of fuel rods or pins, the upper limit of temperature is imposed by the fuel rod or fuel pin cladding material employed. In order to adequately protect the reactor core against excessive temperatures, it is necessary to examine the temperature of the "hottest" fuel pin or the hottest coolant channel between adjacent fuel pins of the core since demage will first occur in the hottest fuel pin. Thus, the hottest pin or channel becomes the limiting pin or channel for the reactor core.
As is well known, heat is generated in a reactor by the fission process in the fuel material. The fission process, however, produces not only heat but radioacitve isotopes which are potentially harmful and which must be prevented from escaping to the environment. To this end, the fuel is clad with a material which retains the fission products. In order to prevent clad overheating, in the interest of precluding the release of the fission products which occur on clad damage or failure, a coolant is circulated through the reactor core. Heat transferred to the circulating coolant from the fuel elements is extracted therefrom in the form of usable energy downstream of the reactor core in a steam generator. Thus, for example, in a pressurized water reactor system the water flowing through the core is kept under pressure and is pumped on the tube side of a steam generator where its heat is transferred to the water on the shell side of the generator. The water on the shell side of the steam generator is under lower pressure and thus thermal energy transfer causes the secondary water to boil and the stream so generated is employed to drive a turbine.
To summarize, in the design and operation of a nuclear reactor, the basic objective of removing heat from the fuel must be obtained without allowing the temperature of the fuel cladding of the limiting fuel pin to rise to such a degree that the clad will fail.
As the coolant circulates through the reactor core heat will be transferred thereto either through sub-cooled convection, often referred to as film conduction, or through nucleate boiling. Nucleate boiling occurs at higher levels of heat flux and is the preferred mode of operation since it permits more energy to be transferred to the coolant thereby permitting the reactor to be operated at a higher level of efficiency. Nucleate boiling is characterized by the formation of steam bubbles at nucleation sites on the heat transfer surfaces. These bubbles break away from the surface and are carried into the main coolant stream. If the bulk coolant enthalpy is below saturation, the steam bubbles collapse with no net vapor formation in the channel. This phenomenon is called sub-cooled boiling or local boiling. If the bulk fluid enthalpy is at or above the enthalpy of saturated liquid, the steam bubbles do not collapse and the coolant is said to be in bulk boiling.
If the heat flux is increased to a sufficiently high value, the bubbles formed on the heat transfer surface during nucleate boiling are formed at such a high rate that they can not be carried away as rapidly as they are formed. The bubbles then tend to coalesce on the heat transfer surface and form a vapor blanket or film. This film imposes a high resistance to heat transfer and the temperature drop across the film can become very large even though there is no further increase in heat flux. This transition from nucleate boiling to film boiling is called "departure from nucleate boiling", hereinafter referred to as DNB, and the value of the heat flux at which DNB occurs is called the "DNB heat flux" in a pressurized water reactor and the "critical heat flux" in a boiling water reactor. Similarly, if the quantity of steam per coolant volume becomes too great a condition known as "excessive void fraction" may occur. Excessive void fraction may result in flow instabilities or a decrease in the heat transfer coefficient from the cladding to the coolant.
Since clad damage is likely to occur because of a decrease in heat transfer coefficient and the accompanying higher clad temperature which may result when DNB or excessive void fraction occurs, the onset of these conditions must be sensed or predicted and corrective action in the form of a reduction in fission rate promptly instituted. Restated, in reactor operation DNB must be prevented since the concurrent reduction in clad strength as temperature increases can lead to a clad failure because of the external coolant pressure or because of the internal fission gas pressures in the fuel rod. One way of monitoring DNB in a reactor is to generate an index or correlation which indicates the reactor condition with respect to the probability of the occurrence of DNB. For a theoretical discussion of the prediction of the onset of DNB, reference may be had to the article "Prediction Of DNB For An Axially Non-Uniform Heat Flux Distribution" by L. S. Tong which appeared in the Journal Of Nuclear Energy, 21:241, 1967.
The ratio of the heat flux necessary to achieve DNB at specific local coolant conditions to the actual local heat flux is known in the art as the departure from nucleate boiling ratio (DNBR) or the critical heat flux ratio. The two correlations, DNBR and critical heat flux ratio, are based upon slightly differing statistical derivations such that the critical values of DNBR and critical heat flux ratio are defined to be 1.3 and 1 respectively. These are the statistically established limiting values above which DNB has a very small probability of occurring. As employed herein, in the interest of facilitating understanding of the invention, DNBR will be used to describe both correlations. Thus, for the purposes of this discussion and description, DNBR shall mean both the Tong W-3 correlation for departure from nucleate boiling ratio and the critical heat flux ratio correlation.
It is known that DNB and excessive coolant void fraction occur as functions of the reactor operating parameters of heat flux or power distribution, primary coolant mass flow rate, primary coolant pressure and primary coolant temperature. In order to prevent an excessive coolant void fraction or DNB, also called "burn-out" or "boiling crisis", reactor protective systems must be designed to insure that reactor operation is rapidly curtailed, a condition known in the art as "reactor trip" or "reactor scram", before the combination of conditions commensurate with DNB or excessive coolant void fraction can exist. Departure from nucleate boiling and DNB ratio may be expressed for one fuel pin or channel as: EQU DNBR = f [ Q, T.sub.C, P, W, F.sub.r, F.sub.2 (Z), EQU T.sub.AZ ] (1)
where
Q = core power in percent of full power PA1 T.sub.C = coolant inlet temperature PA1 P = primary or reactor coolant system pressure PA1 W = coolant mass flow rate PA1 F.sub.r = integral radial peaking factor PA1 F.sub.z (Z) = axial power distribution in the pin which has the integral radial power peaking factor PA1 T.sub.AZ = azimuthal tilt magnitude (the azimuthal component of power distribution) which is a measure of side to side xenon tilt.
In computing DNBR, core power in percent of full power may be determined in a manner similar to that disclosed in U.S. Pat. No. 3,752,735 entitled "Instrumentation for Nuclear Reactor" and assigned to the assignee of the present invention. Integral radial power peaking factor is defined as the maximum ratio of power generated in any fuel pin in the core to the average fuel pin power in the absence of aximuthal flux tilt. Axial power distribution is defined for each fuel pin as a curve of local pin power density versus axial distance up the pin divided by the total power generated in the pin.
Solutions to the problem of protective system design assume that primary coolant mass flow rate, integral radial peaking factor and azimuthal tilt magnitude are maintained within predetermined limits during numerous events which necessitate a reactor trip to prevent the DNBR or coolant void fraction limits from being exceeded. Prior art approaches to protective system design have also assumed that the axial distribution of power in the reactor core was maintained within the limits of its normal operating envelope. For a full disclosure of a prior art thermal margin protection system based on the preceding assumptions, reference may be had to U.S. Pat. No. 3,791,922 entitled "Thermal Margin for a Nuclear Reactor Protection System" which is assigned to the same assignee as the present invention. U.S. Pat. No. 3,791,922 contains a detailed discussion of the means by which the locus of points at which a DNB or excessive coolant void fraction thermal limit will occur and the disclosure of said copending application is incorporated herein by reference.
Heretofore the prior art, including the technique and apparatus of referenced U.S. Pat. No. 3,791,922 has maintained core protection through means and methods which have been unduly conservative and thus have sacrificed plant operating margins. The assumption that certain operational parameters, and particularly axial power distribution, were either constants held at their design values or were variables which varied only within their allowed envelopes, has precluded reactor operation at power levels approaching the optimum for the existing conditions. The economic penalty imposed by unduly conservative safety system design is particularly apparent in the case of very large and high power reactors.